Flow Field and Thermal Analysis of the Divertor Target Plate for HL-2A Tokamak

In the initial phase of the physics experiment, the double-null divertor plates used consist of graphite armor tiles, Mo-alloy intermediate layers and Cu-alloy coolant tubes. In the later operating phase, tungsten will be used as armor tiles. A multi-physical field numerical analysis method is used in this paper. Its analysis model reflects more realistically the real divertor structure than other models. Two-dimensional （2D） and three-dimensional （3D） fluid flow field, temperature distribution and thermal stress analyses of the divertor plates are carried out by the ANSYS code. During the physics experimental phase with a heat flux of 1 MW/m2, a coolant velocity of 5.48 m/s, and a thermal stress of 750 kg/cm2, the graphite armor tiles successfully meet the requirements of temperature, thermal stress and sputtering erosion. The tungsten armor will be considered as a second candidate. The result of simulation can be used for upgrading the design parameters of the HL-2A poloidal divertor.     

3D finite element analysis of a nuclear fuel rod with gap elements between the pellet and the cladding

Nuclear fuel rods which comprises an important component of a nuclear power plant are composed of nuclear fuel and cladding. Simulating the nuclear fuel rod using a computer program is the universal method to verify its safety. The computer program used for this is called the fuel performance code. The main objective of this study is to simulate the nuclear fuel rod behavior considering the gap conductance using three-dimensional gap elements. Gap elements are used because, unlike other methods, this approach does not require special methods or other variables such as the Lagrange multiplier. In this work, a nuclear fuel rod has been simulated and the results are compared with the experimental results.     

Some insights into the role of axial gas flow in fuel rod behaviour during the LOCA based on Halden tests and calculations with the FALCON-PSI code

A model for axial gas flow in a fuel rod during the LOCA is integrated into the FRELAX model that deals with the thermal behaviour and fuel relocation in the fuel rods of the Halden LOCA test series. The first verification was carried out using the experimental data for the inner pressure during the gas outflow after cladding rupture in tests 3, 4 and 5. Furthermore, the modified FRELAX model is implicitly coupled to the FALCON fuel behaviour code.The analysis with the new methodology shows that the dynamics of axial gas-flow along the rod and through the cladding rupture can have a strong influence on the fuel rod behaviour. Specifically, a delayed axial gas redistribution during the heat-up phase of the LOCA can result in a drop of local pressure in the ballooned area, which is eventually able to affect the cladding burst. The results of the new model seem to be useful when analysing some of the Halden LOCA tests (showing considerable fuel relocation) and selected cases of LOCA in full-length fuel rods. While the short rods used in the Halden tests only show a very small effect of the delayed gas redistribution during the clad ballooning, such an effect is predicted to be significant in the full-scale rods - with a power peak located sufficiently away from the plenum - resulting in a considerable delay of the predicted moment of cladding rupture.     

Analysis of thermal problems in development of spent fuel transport cask for research reactors

A packaging for the transport of irradiated fuel from research reactors was designed by a group of researchers to improve the capability in the management of spent fuel elements from the reactors operated in the region. Two half scale models for MTR fuel were constructed and tested so far and a third one for both MTR and TRIGA fuels will be constructed and tested next. Four test campaigns have been carried out, covering both normal and hypothetical accident conditions of transportation. The thermal test is part of the requirements for the qualification of transportation packages for nuclear reactors spent fuel elements. In this paper, both the numerical modelling and experimental thermal tests performed are presented and discussed. The cask is briefly described as well as the finite element model developed and the main adopted hypotheses for the thermal phenomena. The results of both numerical runs and experimental tests are discussed as a tool to validate the thermal modelling. The impact limiters, attached to the cask for protection, were not modelled.     

Assessment of fuel rod performance codes under ramp scenarios investigated within the SCIP project

The behavior of advanced cladding materials under challenging conditions needs to be fully characterized, understood and modeled. This paper assesses the current predictability of fuel performance codes under loadings expected from pellet-clad mechanical interactions. A set of scenarios experimentally characterized within the SCIP project, were chosen so that a variety of materials and ramp power sequences could be examined.Four codes have been used in this study: ALCYONE V1.1, FALCON-PSI, FRAPCON-3 v3.3 and STAV7.3. Their predictions have been compared to data in terms of cladding oxidation, diameters and elongation. Predictability of clad oxidation was certainly scattered and while some codes showed reasonable accuracy, other results were notably deviated. As for diameters, most of the codes were capable of qualitatively capturing the axial profile, and showed consistency between diameters and hoop stress and strain predictions. Elongation estimates were generally poor, and were rather far from measurements in most cases (even the trends observed just vaguely followed by the codes).The results reported have been discussed in the light of the set of individual hypotheses and approximations made by modelers and codes regarding both boundary conditions (i.e., power histories, inlet coolant temperature, refabrication, etc.) and fuel and clad characterization (i.e., densification, rim porosity, materials properties, etc.). Additionally, code-to-code comparisons of some key variables (i.e., fuel temperature, contact pressure, hoop and axial stresses, etc.) highlighted systematic tendencies of the codes and supported the observations made.     

穿透裂纹临界泄漏率分析程序的研制与验证

建立考虑裂纹形态参数影响的周向穿透裂纹临界泄漏率的计算模型,以此为基础编制计算程序PC-Leakflow2。介绍程序的计算流程及求解方法,对影响裂纹临界泄漏率的各个输入参数进行敏感性分析,用文献中的临界泄漏率试验数据对PC-Leakflow2程序的计算结果进行验证。用PC-Leakflow2程序和经典的临界流模型对相同的例题进行计算,计算结果表明:临界泄漏率的大小受裂纹形态参数的影响较强;经典的临界流模型会显著地高估紧密裂纹的临界泄漏率。     

Analysis of surface damage produced by pulsed laser ablation on metal Al and semiconductor Si

The suraface morphological changesd produced by Nd:YAG pulsed laser ablation of metal Al and semiconductor Si were carefully examined and analyzed by using scanning elkectron microscope.The formation mechanism of the droplets was discussed.and the reasons for formation of the microcracks on the laser irradiated area of the target surface were analyzed by calculating the thermal stress,the vapor pressure and the shock pressure induced by the laser supported detonation.     

某核电厂泵房基坑边坡开挖支护数值模拟分析

本研究对象为某核电厂取水泵房基坑边坡。根据边坡设计的开挖支护方案,采用ANSYS有限元数值模拟软件,对边坡的开挖支护过程进行了仿真数值模拟。模拟结果表明,在边坡采用深搅桩加固、支护桩、预应力锚索等工程措施后,上台边坡和基坑边坡开挖后引起的位移和应力均较小,边坡处于稳定状态。     

用于快堆非能动停堆系统的磁性连接对的设计优化与完整性综合评价

按照ASME规范要求,完成了用于快堆非能动停堆系统的磁性连接对的设计;对工况进行了分析,总结出了磁性连接对在正常运行工况下的载荷,并利用大型有限元分析软件ANSYS对连接对进行计算,得到正常运行工况下的应力值。按照ASME规范,对磁性连接对在正常运行工况下进行了应力强度分析和评定及疲劳评定。结果显示,磁性连接对在正常运行工况下的一次和二次应力强度及疲劳评定均满足ASME规范的要求。     

熔盐自然循环回路热损失功率实验及计算

熔盐自然循环回路是为研究熔盐的自然循环特性,支持先进熔盐堆非能动安全系统设计而建造的实验装置。熔盐在回路中的热量损失对于自然循环的建立和保持具有重要的影响。本文以熔盐储罐为代表部件,通过实验得到了其热量损失的数据,并利用数值模拟的方法,计算了不同温度下储罐各部分的热损失,分析了储罐热损失规律,拟合得到了热损失随温度及时间的变化关系。对比熔盐在不同温度下热损失的实验值和计算值,发现两者吻合良好,相对误差均小于10%。分析结果表明,内插式电加热器是储罐主要热损失途径之一,并导致了熔盐的温度分层。     

An alternative solution of the neutron diffusion equation in cylindrical symmetry

Alternative analytical solutions of the neutron diffusion equation for both infinite and finite cylinders of fissile material are formulated using the homotopy perturbation method. Zero flux boundary conditions are investigated on boundary as well as on extrapolated boundary. Numerical results are provided for one-speed fast neutrons in ²³⁵U. The results reveal that the homotopy perturbation method provides an accurate alternative to the Bessel function based solutions for these geometries.     

理想化的实际反应堆内热中子有效增殖系数的定义及表达式的讨论

目前教科书中介绍的反应堆热中子有效增殖系数keff,是对无源中子的反应堆内的中子变化情况的准确定义及相应表达式的准确介绍,它能准确解释并描述中子在反应堆内六种物理过程中的变化情况。但用它作为一个统一的定义及表达式来描述并解释和计算有源中子存在的实际反应堆时,对于部分情况它既不能准确、清楚,又不能正确解释相应的物理过程,它的表达式也不能作为一个统一的表达式,按照它的定义计算得到相应的结果。且,目前,国内外工程研究人员还没有给出过实际反应堆内有源中子存在情况下的热中子有效增殖系数的定义及表达式。因此,特撰写此文,对考虑了源中子后的实际反应堆的热中子有效增殖系数,给出一个正确且准确的定义及相应表达式。     

New modelling of the U-O-Zr phase diagram in the hyper-stoichiometric region and consequences for the fuel rod liquefaction in oxidising conditions

Liquidus and solidus temperatures were recently re-measured in the UO_2+x composition domain by [D. Manara, C. Ronchi, M. Sheindlin, M. Lewis, M. Brykin, J. Nucl. Mater. 342 (2005) 148]. The main difference with the Latta and Fryxell’s data [R.E. Latta, R.E. Fryxell, J. Nucl. Mater. 35 (1970) 195] data is that the Manara’s transition temperatures were accurately determined using a self-crucible technique while the former data were obtained in a W crucible and then suspected of crucible contamination. According to these recent data, a new thermodynamic modelling of U-O phase diagram is here presented and introduced in the European NUCLEA thermodynamic database for corium applications. An important consequence of this new optimisation for safety applications is that a liquid phase may appear in the O-UO₂-ZrO₂ composition domain of the U-O-Zr phase diagram at 2600 K at atmospheric pressure (this temperature decreasing with increase of pressure, about 2500 K at 2 atm.). These temperatures can be associated with the temperature at which the fuel assembly could loose its integrity in oxidising conditions and then with what was observed in some of the VERCORS tests where fuel collapse was detected in the temperature range of 2400-2600 K (and quite differently from reducing test conditions) or in the PHEBUS tests where indications of early fuel collapse at 2500-2600 K were identified.     

NSRL前端活动水冷光屏的结构及热分析

活动水冷光屏是光束线前端的重要部件，主要用来吸收高强度的同步辐射功率、承受和释放热负载，保护其下游位置仪器设备免遭热载破坏。介绍中国科学技术大学国家同步辐射实验室(NSRL)的合肥同步辐射光源(HLS)光束线前端活动水冷光屏的结构，利用有限元分析程序ANSYS5．5进行热力学分析，确认其合理性。     

Numerical solutions for the kinematic model of pebble flow velocity profiles and its applications in pebble-bed nuclear reactor

The modular pebble-bed nuclear reactor (PBR) is a candidate Generation IV reactor being developed. The pebble flow in the very slow draining of fuel pebbles draws attention for its implications on core physical design and reactor physics analysis. One of the effective and simplified methods to address this problem is the kinematic model which is based on continuous theory to derive a diffusion equation for vertical velocity. This paper investigates the appropriate numerical solutions for the kinematic model of pebble flow velocity profiles in PBR geometry. Our method is based on a previously proposed transformed Cartesian coordinates and uses the implicit Crank–Nicholson integration scheme with two different treatments of the boundary conditions. Validations show that this numerical solution gives preferable agreements with the experimental results in the reference. Finally, the simulated velocity profiles are applied in the investigation of two pebble burnup-related issues, which are the pebble residence time prediction and the channel scheme in realistic high-temperature reactor pebble-bed modules reactor core geometry.     

带热套管的T型接管内流动换热的数值模拟和实验研究

为了分析核反应堆冷却剂系统中带热套管T型接管内由于注入非等横向射流导致的构件热冲击状况，本文应用计算流体力学商用软件FLUENT5．3进行了紊流流动换热的数值模拟，分析了主管及接管与热套间环腔内的流动换热特性，针对套管上开有通流小孔，并采用凸台支撑的热套管结构形式，模拟了射流与主流流速比为0．05及0．5两种典型工程，传热实验，研究了主管及接管内壁近壁区域的传热特性，并讨论了热套管尺寸变化对接管热冲击的影响，结果表明，数值模拟与实验数据吻合良好，热套管对构件的热保护程度与热套管结构形式及流速比密切相关，适当减小流速比有利于改善构件热应力状况。     

Analysis of core physics experiments on irradiated BWR MOX fuel in the REBUS program

Critical experiments were performed in the REBUS program on a core loaded with a test bundle including 16 irradiated BWR-type MOX rods of average burnup of 61 GWd/t. The experimental data were analyzed using diffusion, transport, and continuous-energy Monte Carlo calculation codes coupled with nuclear data libraries based on JENDL-3.2 or JENDL-3.3. Biases in effective multiplication factors of the critical cores were ?1.0%Δk for the diffusion calculations (JENDL-3.2), ?0.3%Δk for the transport calculations (JENDL-3.3), and 0.2%Δk for the Monte Carlo calculations (JENDL-3.2). The measured core fission rate and co-activation rate distributions were generally well reproduced using the three types of calculations. The burnup reactivity determined using the measured water level reactivity coefficients was ?2.41 ± 0.08%Δk/kk’, which also agreed with the results of the three type of calculations within the measurement and calculation errors. The most probable isotopic inventories in the irradiated MOX rods was tentatively obtained by using the ratios of the calculation to chemical assay data on a pellet sample, and the burnup reactivity was reanalyzed to split the calculation error into those due to the inventory and reactivity calculations. This approach showed that the inventory calculation error compensated the reactivity calculation error.     

Thermal-hydraulic performance analysis of a subchannel with square and triangle fuel rod arrangements using the entropy generation approach

The present paper discusses entropy generation in fully developed turbulent flows through a subchannel,arranged in square and triangle arrays. Entropy generation is due to contribution of both heat transfer and pressure drop. Our main objective is to study the effect of key parameters such as spacer grid, fuel rod power distribution,Reynolds number Re, dimensionless heat power ω, lengthto-fuel-diameter ratio λ, and pitch-to-diameter ratio ξ on subchannel entropy generation. The analysis explicitly shows the contribution of heat transfer and pressure drop to the total entropy generation. An analytical formulation is introduced to total entropy generation for situations with uniform and sinusoidal rod power distribution. It is concluded that power distribution affects entropy generation.A smoother power profile leads to less entropy generation.The entropy generation of square rod array bundles is more efficient than that of triangular rod arrays, and spacer grids generate more entropy.     

Post-facta analysis of the TMI accident (II): Analysis of fuel rod behavior and core damage estimation by use of TOODEE2-J

By use of the TOODEE2-J computer program, an analysis was carried out of the fuel rod behavior, and core damage was estimated for the TMI-2 reactor during the first three hours of the accident on March 28, 1979. The boundary conditions (e.g. core mixture level, steam flow rate and core inlet flow) are based on a thermal-hydraulic analysis by the RELAP4/MOD6/U4/J2 computer program. The calculated results suggest that bursting of almost all rods except peripheral low-powered rods occurred, and that a large part of the zircaloy cladding exceeded the eutectic temperature to form a liquid phase of Zr---U---O. A total of 43.5% of the zircaloy in the fueled part of the core had been converted to zirconium-dioxide by three hours into the accidient, and major damage to the fuel rods had also occured by then.     

Calculation of fuel burnup and radionuclide inventory in the Syrian miniature neutron source reactor using the GETERA code

Calculations of the fuel burnup and radionuclide inventory in the Syrian miniature neutron source reactor (MNSR) after 10 years (the reactor core expected life) of the reactor operation time are presented in this paper using the GETERA code. The code is used to calculate the fuel group constants and the infinite multiplication factor versus the reactor operating time for 10, 20, and 30 kW operating power levels. The amounts of uranium burntup and plutonium produced in the reactor core, the concentrations and radionuclides of the most important fission products and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core were calculated using the GETERA code as well. It is found that the GETERA code is better than the WIMSD4 code for the fuel burnup calculation in the MNSR reactor since it is newer, has a bigger library of isotopes, and is more accurate.